System for extraction of tritium from liquid metal coolants

ABSTRACT

A method for removing tritium from liquid lithium includes mixing the liquid lithium containing trace amounts of tritium with a molten salt and forming a salt of lithium and tritium. The method also includes separating the liquid lithium from the salt of lithium and tritium and circulating the molten salt in an electrolyzer to form molecular tritium. The method further includes bubbling an inert gas through the electrolyzer to remove the molecular tritium and circulating the argon and removed molecular tritium in a titanium getter to recover the tritium.

CROSS-REFERENCES TO RELATED APPLICATIONS

This application claims priority to U.S. Provisional Patent ApplicationNo. 61/722,569, filed Nov. 5, 2012, and entitled: “System for Extractionof Tritium from Liquid Metal Coolants,” the disclosure of which ishereby incorporated by reference in its entirety for all purposes.

STATEMENT AS TO RIGHTS TO INVENTIONS MADE UNDER FEDERALLY SPONSOREDRESEARCH OR DEVELOPMENT

The United States Government has rights in this invention pursuant toContract No. DE-AC52-07NA27344 between the United States Department ofEnergy and Lawrence Livermore National Security, LLC for the operationof Lawrence Livermore National Laboratory.

BACKGROUND OF THE INVENTION

The National Ignition Facility (NIF), the world's largest and mostenergetic laser system, is operational at Lawrence Livermore NationalLaboratory (LLNL) in Livermore, Calif. One goal of operation of the NIFis to demonstrate fusion ignition for the first time in the laboratory.Initial experiments are calculated to produce yields of the order of 20MJ from an ignited, self-propagating fusion burn wave. The capability ofthe facility is such that yields of up to 150-200 MJ could ultimately beobtained. The NIF is designed as a research instrument, one in whichsingle “shots” on deuterium-tritium containing targets are performed forresearch. A description of the NIF can be found in Moses et al, FusionScience and Technology, volume 60, pp 11-16 (2011) and referencestherein.

There is a rapidly growing need for power, and especially for cleanpower. At LLNL a project known as Laser Inertial-confinement FusionEnergy, (often referred to herein as “LIFE”) is working towardintroduction of fusion based electric power plants into the U.S. economybefore 2030, and in a pre-commercial plant format before that. LIFEtechnology offers a pathway for the expansion of carbon-free poweraround the world. It will provide clean carbon-free energy in a safe andsustainable manner without risk of nuclear proliferation.

One challenge with respect to LIFE, as well as any technology forgenerating electrical power to be distributed to large numbers ofconsumers, is reliability. Consumers expect to have extraordinarily highreliability in their electric power supply. The result is that utilitiesthat provide that electrical power maintain their facilities to assurethe required high reliability. Thus, among the challenges with respectto fusion power, is to provide mechanisms by which a reliable long-livedfusion chamber can be provided in which the fusion reactions occur, yetwhich can be maintained or replaced when necessary with minimal downtimefor the overall power plant.

Despite the progress made in the design of fusion based electric powerplants, there is a need in the art for improved methods of extractingtritium from liquid coolants.

SUMMARY OF THE INVENTION

The present invention relates generally to methods and systems forextracting tritium from lithium-based coolants. More particularly,embodiments of the present invention provide methods and systems forextracting tritium from lithium liquids that includes trace amounts oflithium. Although embodiments of the present invention are discussed interms of lithium-based coolants, other liquid metal coolants areincluded within the scope of the present invention.

According to an embodiment of the present invention, a method forremoving tritium from liquid lithium is provided. The method includesmixing the liquid lithium containing trace amounts of tritium with amolten salt and forming a salt of lithium and tritium. The method alsoincludes separating the liquid lithium from the salt of lithium andtritium and circulating the molten salt in an electrolyzer to formmolecular tritium. The method further includes bubbling an inert gasthrough the electrolyzer to remove the molecular tritium and circulatingthe argon and removed molecular tritium in a titanium getter to recoverthe tritium.

According to another embodiment of the present invention, a system forrecovering tritium from a lithium coolant is provided. The systemincludes a lithium coolant circuit. The lithium coolant in the circuitincludes trace amounts of tritium. The system also includes a pluralityof contactors coupled to the lithium coolant circuit. The plurality ofcontactors are operable to mix the lithium coolant with a molten saltand form a salt of lithium and tritium. The system further includes anelectrolysis unit coupled to one or more of the plurality of contactors,an inert gas source operable to bubble an inert gas through theelectrolysis unit, and a getter coupled to the electrolysis unit andoperable to recover the tritium.

Numerous benefits are achieved by way of the present invention overconventional techniques. For example, embodiments of the presentinvention provide methods and systems for extracting tritium fromlithium coolant fluids that are more efficient than conventionaltechniques. These and other embodiments of the invention along with manyof its advantages and features are described in more detail inconjunction with the text below and attached figures.

BRIEF DESCRIPTION OF THE DRAWINGS

FIG. 1 is a simplified schematic diagram illustrating a 1.2 GW (thermal)power plant based on the LIFE engine.

FIG. 2 is a simplified schematic diagram of centrifugal contactorsconnected in series.

FIG. 3 is a simplified plot of the liquid lithium flow processed torecover tritium as a function of extraction efficiency of centrifugalcontactors.

FIG. 4 is a simplified plot showing the liquid lithium fraction that isprocessed to maintain 0.1 wppm steady state tritium concentration.

FIG. 5 is a plot showing the number of contactors as a function of thecentrifugal contactor extraction efficiency.

FIG. 6 is a simplified schematic diagram of a tritium recovery system bymolten salt extraction according to an embodiment of the presentinvention.

FIG. 7 is a simplified schematic diagram of a tritium recovery systemutilizing redundancy.

FIG. 8 is a simplified flowchart illustrating a method of extractingtritium according to an embodiment of the present invention.

DETAILED DESCRIPTION OF SPECIFIC EMBODIMENTS

Liquid lithium is considered as the primary coolant and breeder materialin the design of the Laser Inertial Fusion Energy (LIFE) engine. Thiswork presents the analysis of a tritium recovery system from the blanketliquid lithium for the LIFE engine and is based on a molten saltextraction technology. The goal is to size the primary components of thesystem. The dimensions of the centrifugal contactors, electrolyzer, andgetter are estimated as a function of the desired tritium steady stateinventory and the component efficiencies. Results show that a relativelycompact system can recover tritium at the needed rate to maintain a lowtritium inventory.

The LIFE Engine is a laser-based energy system. Liquid lithium is beingconsidered as the blanket cooling and breeding material for the LIFEengine due to its many virtues: low density, high thermal capacity, goodthermal conductivity, high potential to breed tritium due to its highneutron capture cross-section, and high affinity for hydrogen isotopesthat minimizes tritium permeability. High hydrogen affinity, however,complicates tritium recovery while operational safety demands a lowtritium inventory. Contact with water is avoided at all times.

The tritium recovery system is designed to keep the tritium steady stateinventory of the LIFE engine power plant at less than about 50 g. Theamount of liquid lithium in LIFE pilot plant is ˜1000 tons. If we have˜1000 m³ of liquid lithium in our 1.2 GW(th) LIFE plant we need a steadysate tritium concentration of 0.1 wppm. If our tritium recovery systemhas 90% efficiency then we need a system that can process ˜1% of thelithium flow to decrease tritium concentration from 0.1 wppm to 0.01wppm.

FIG. 1 is a simplified schematic diagram illustrating a 1.2 GW (thermal)power plant based on the LIFE engine. The design of the pilot LIFEengine power plant includes four loops, each with an intermediate heatexchanger, steam generator, and reheater. Thus one tritium recoverysystem processes ˜7 kg/s of lithium flow to reduce tritium concentrationfrom 0.1 wppm to 0.01 wppm.

FIG. 1 illustrates two of the four loops. First, liquid lithiumtransfers energy from the fusion chamber 110 to the intermediate heatexchanger 112 where the thermal energy of the liquid lithium istransferred to the molten salt. The molten salt in the secondary loop(also referred to as an intermediate loop) carries thermal energy fromthe intermediate heat exchanger 112 to the steam generator and reheater114. The steam then goes to the Rankin cycle turbines. Each of the fourloops will also have a tritium recovery system by molten salt extraction(labeled “Lithium Processing” in FIG. 1).

Several options have been proposed for tritium recovery from liquidlithium. These include use of a permeable window, a gettering process, acold trap, distillation, molten salt, and gettering and molten saltcombined. Because the molten salt method is based on liquid diffusion atrelatively low temperature (˜500° C.) it results in a relatively compactsystem with relatively low energy consumption. For that reason, andothers discussed below, the molten salt extraction is utilized in theembodiments described herein.

FIG. 8 is a simplified flowchart illustrating a method of extractingtritium according to an embodiment of the present invention. Asillustrated in FIG. 8, the molten salt extraction method includes foursteps in this embodiment. The method includes mixing a lithium fluidcontaining trace amounts of tritium with a molten salt (e.g., lithiumhalides, for example, including LiF, LiCl, LiBr, or combinationsthereof) in a centrifugal contactor (810). In other embodiments, one ormore sodium salts can be utilized as the molten salt. The intimatecontact between the molten salt and the trace amounts of tritium in thelithium fluid preferentially extracts the LiT into the salt phase (i.e.,a salt of lithium and tritium). The lithium fluid and the salt oflithium and tritium are separated in the centrifugal contactor (812).The salt of lithium and tritium is circulated in an electrolyzer and theLiT is oxidized to form T2 (814), which is swept from the salt oflithium and tritium by bubbling an inert gas (e.g., argon) through thesalt (816). Finally, the inert gas is circulated in a titanium getterwhich recovers the tritium from the inert gas (818).

The most important characteristic of the molten salt extraction methodis being free from solid-state diffusion. Liquid phase diffusion isoften faster than solid-state diffusion. However, distillation requiresvery high temperatures. Maroni et al. were the first to propose the useof molten salt extraction for tritium recovery from liquid lithium. Inprevious experiments molten salt had been used to remove impurities fromliquid metals. They observed that some experiments with lithium/hydrogengalvanic cells showed that the lithium hydride formed during dischargewas preferentially extracted into the electrolyte.

As described above, the liquid lithium (also referred to as lithiumliquid) that contains tritium is mixed with lithium halide salts in acentrifugal contactor/separator. It has been shown that the volumetricdistribution coefficient, Dv (defined as the ratio of tritium contentper unit volume in the salt to tritium content per unit volume in thelithium) was between 2 and 4, verifying that the LiT movespreferentially from the liquid lithium to the salt when they are mixed.Then, the salt of lithium and tritium is circulated to an electrolyzerwhere the LiT is oxidized to form T2, which is swept from the salt phaseby bubbling an inert (i.e., noble) gas. The noble gas circulates througha getter that recovers the tritium. It has been estimated that a 20%electrolyzer recovery efficiency is used to capture tritium from themolten salt by measuring the hydrogen isotopes added to the molten saltand the hydrogen concentration in the noble gas coming out of theelectrolyzer.

Embodiments of the present invention provide a molten salt extractionmethod that can be used in a full-scale fusion system. Important issuesthat are addressed are: 1) The need to reduce the halide impurities inthe primary lithium blanket to safe levels from an activationperspective. This depends on how well the centrifugal contactorsseparate the lithium and molten salt after LiT is transferred. 2) Theneed to minimize impurities in the salt that can decrease electrolyzerefficiency. Once again, this depends on how well the contactor-separatorworks.

The following provides an analysis of a complete molten salt extractionsystem to recover tritium from the liquid lithium of the LIFE enginepower plant. We estimate the size of the major components and calculatesystem energy consumption.

Centrifugal Contactors

LiF—LiCl—LiBr (22-31-47 mol %) is the primary choice for molten salt inan embodiment although other mixtures can be utilized. The selection ofthis primary choice was made primarily on melting point consideration:i.e., <450° C. is required. LiF—LiCl—LiBr (22-31-47 mol %) also showedgood stability during testing, demonstrating that the particular ratiodiscussed as the primary choice is not required by the presentinvention. We consider that due to similar properties of the moltenhalides the analysis presented here will be valid if another halide isutilized. Most research has been done with 1:1 lithium to salt volumeratio and no issues have been reported, thus the 1:1 ratio isrecommended in the interest of reducing overall salt inventory in theplant.

FIG. 2 is a simplified schematic diagram of centrifugal contactorsconnected in series. Lithium and salt flow counter-current, mixing andseparating in every centrifugal contactor. Lithium and molten salt flowsare in counter-current.

The centrifugal contactor 1 210 receives salt with no tritium andlithium with low tritium concentration (most tritium in liquid lithiumis found as LiT) because tritium is previously extracted in centrifugalcontactor 2 220 and contactor 3 230. The two flows mix and separate withthe lithium leaving with less tritium and the salt leaving contactor 1210 with a small amount of tritium and entering contactor 2 220 where itmixes with lithium with higher tritium concentration, and so on.Increasing liquid lithium purity utilizes more contactors in series. Itis also possible to connect the centrifugal contactors in parallel. Inthat case, the fraction of lithium flow that needs to be processedincreases because the concentration of tritium will be higher after thesingle contactor stage.

There is a trade-off between the contacting step (centrifugalcontactors) and the salt processing step (electrolyzer). If thecontacting step is larger with high energy consumption then theelectrolyzer can be smaller and use less energy because less salt withhigher tritium concentration will flow from the centrifugal contactorsto the electrolyzer. Recovery efficiency determines the number ofcontactors, and electrolyzer volume is determined by the average currentdensity. The more contactors in series, the less salt volume flows intothe electrolyzer, enabling compact electrolyzer design. Some experimentsdemonstrated 100% hydrogen recuperation efficiency from molten saltelectrolysis. More experimental work is needed, but with the availableinformation it seems better to maximize electrolysis because thecentrifugal contactors are more complex and demand more energy. Thus, insome embodiments, parallel centrifugal contactors are utilized, forexample, for the pilot LIFE engine plant.

Electrolyzer

The electrolysis unit is the least studied component of the system.Recovery efficiency, defined as the weight of the hydrogen recoveredfrom the salt divided by the weight of hydrogen added to the salt, isbelieved to be the most important electrolyzer parameter. Electrolysisexperiments have been conducted to separate hydrogen from molten salt,reporting recovery efficiency ranging from 20% to 100%. The higherefficiency is explained by the use of an argon bubbler that was also theanode of the cell that removed the hydrogen before back mixing occurred.The kinetics of tritium recovery in the contactor units has been studiedbut did not include a study of electrolyzer kinetics. Typically, anassumption of 90% electrolysis process efficiency is used. It has beenshown that applying less than 0.6 V does not release any tritium.Recovery efficiency subsequently increased until reaching a maximum at0.9 V. After that, recovery efficiency decreases for higher voltage.Voltage should not be increased beyond 1.5 V to avoid molten saltdecomposition that occurs at ˜2 V.

In the electrolyzer the LiT in the salt phase is oxidized to becomemolecular tritium (T2). The T2 collection electrode must be designed torecover tritium before it can react back into the salt phase. In anparticular embodiment, a porous electrode enabling surface oxidation ofT2 with simultaneous argon flow through the small electrode orifices isutilized to carry away the tritium before back reaction can take place.

Getter

The noble gas (e.g., Ar)-T2 mixture flows from the electrolyzer to agetter. Even though there might not be many getters specificallydesigned to extract tritium from argon, this component is moreconventional because there are many hydrogen getters, and argon/heliumis unreactive. Several getter materials can be used: titanium, yttrium,and depleted uranium, which is likely the best option.

System Analysis.

For the 1.2 GW (thermal) LIFE engine pilot power plant, tritium breedingis estimated at Rb=200 g/d. The total volume of liquid lithium (blanket,pipes, pumps, and intermediate heat exchangers) is estimated to be 100m³. At density pLi=480 kg/m³ at T=500° C., the total lithium mass is 480tons. The LIFE engine power plant has the goal of keeping the steadystate tritium inventory <50 g. Tritium steady state concentration in thelithium loop is therefore ˜0.1 wppm. If we used the values of DV=2.9(measured) and ε=90%, we can plot the liquid lithium flow that needs tobe processed to recover the tritium as a function of the extractionefficiency of the centrifugal contactors (ii). FIG. 3 is a simplifiedplot of the liquid lithium flow processed to recover tritium as afunction of extraction efficiency of centrifugal contactors. Extractionefficiency is difficult to calculate or estimate because it depends onspecific geometry and operating conditions of the centrifugalcontactors.

In one implementation, one volume of liquid lithium is mixed with threevolumes of molten salt, resulting in a total mixed flow of 210 m₃/h. Ifwe use the 45 m3/h (200 gpm) centrifugal contactor we need 5 sets ofthree contactor per loop, or 15 contactor per loop, 60 contactors total.

FIG. 4 is a simplified plot showing the liquid lithium fraction that isprocessed to maintain 0.1 wppm steady state tritium concentration. Toestimate the number of centrifugal contactors we selected a commercialproduct that has a maximum throughput of 757 liters per minute (200gpm), with a footprint of 152 cm×152 cm (60″×60″) and height 163 cm(64″). Using an estimate for the extraction efficiency, it is possibleto calculate the number of contactors needed for the whole LIFE engineplant as a function of the centrifugal contactor extraction efficiency.FIG. 5 is a plot showing the number of contactors as a function of thecentrifugal contactor extraction efficiency.

If we conservatively assume n=0.4, the total number of centrifugalcontactors is 16, or 4 contactors in parallel per loop. FIG. 6 is asimplified schematic diagram of a tritium recovery system by molten saltextraction according to an embodiment of the present invention. Asillustrated in FIG. 6, with four centrifugal contactors (CentrifugalContactors 1-4) in parallel, the electrolyzer 612, the getter 614, andthe lithium fluid circuit 620 and the molten salt circuit 622. In thisexample, a depleted uranium getter is utilized. Additionally, althoughargon is used as the noble gas, this is not required by the presentinvention. As previously described, the molten salt flows from each ofthe centrifugal contactors to an electrolyzer 612 where the T− isoxidized to form T2, and a stream of argon sweeps the molecules of T2before back reaction occurs. The argon stream with tritium (and possiblyother impurities) goes through a getter unit 614 where depleted uraniumabsorbs the tritium, which is later released by heating during theregeneration stage.

The tritium recovery system by molten salt extraction illustrated inFIG. 6 is suitable for use with the 1.2 GW (thermal) LIFE engine powerplant or other suitable sources that generate tritium.

FIG. 7 is a simplified schematic diagram of a tritium recovery systemutilizing redundancy. Examples of the redundant systems include theredundant contactors (Centrifugal Contactor 1 Redundant, CentrifugalContactor mostly salt Redundant, and Centrifugal Contactor mostlylithium Redundant), redundant electrolysis tank (Electrolysis TankRedundant) and the redundant getters (Getters Redundant). One ofordinary skill in the art would recognize many variations,modifications, and alternatives.

It is also understood that the examples and embodiments described hereinare for illustrative purposes only and that various modifications orchanges in light thereof will be suggested to persons skilled in the artand are to be included within the spirit and purview of this applicationand scope of the appended claims.

What is claimed is:
 1. A method for removing tritium from liquidlithium, the method comprising: mixing the liquid lithium containingtrace amounts of tritium with a molten salt; forming a salt of lithiumand tritium; separating the liquid lithium from the salt of lithium andtritium; circulating the molten salt in an electrolyzer to formmolecular tritium; bubbling an inert gas through the electrolyzer toremove the molecular tritium; and circulating the argon and removedmolecular tritium in a titanium getter to recover the tritium.
 2. Themethod of claim 1 wherein the molten salt includes lithium halides. 3.The method of claim 2 wherein the halides comprise at least one offluorine, chlorine, or bromine.
 4. The method of claim 1 wherein theliquid lithium contains trace amounts of tritium.
 5. The method of claim1 wherein mixing the liquid lithium with the molten salt and separatingthe liquid lithium from the molten salt are performed in a centrifugalcontactor.
 6. The method of claim 1 wherein the inert gas comprises atleast one of helium or argon.
 7. A system for recovering tritium from alithium coolant, the system comprising: a lithium coolant circuit,wherein the lithium coolant in the circuit includes trace amounts oftritium; a plurality of contactors coupled to the lithium coolantcircuit, wherein the plurality of contactors are operable to mix thelithium coolant with a molten salt and form a salt of lithium andtritium; an electrolysis unit coupled to one or more of the plurality ofcontactors; an inert gas source operable to bubble an inert gas throughthe electrolysis unit; and a getter coupled to the electrolysis unit andoperable to recover the tritium.
 8. The system of claim 7 wherein themolten salt comprises at least one of LiF, LiCL, or LiBr.
 9. The systemof claim 7 wherein the plurality of contactors comprise centrifugalcontactors arranged in parallel.
 10. The system of claim 7 wherein thegetter comprises at least one of a titanium or depleted uranium.
 11. Thesystem of claim 7 wherein the inert gas comprises at least one of heliumor argon.